The present invention relates to a method of efficiently removing Cs and Sr (hereinafter collectively referred to as "exothermic elements") from a high-level radioactive liquid waste (hereinafter referred to simply as "high-level liquid waste") generated in the step of reprocessing spent nuclear fuels, prior to solidification, such as vitrification or glassification, of the high-level liquid waste. The content of such exothermic elements in a vitrified waste is important for safe storage of the vitrified waste.
A spent nuclear fuel generated in, for example, nuclear power stations contains plutonium and uranium which are fissionable substances. In order to reuse these elements, the spent nuclear fuel is reprocessed to thereby separate and recover plutonium and uranium. In this process, a high-level liquid waste is produced which comprises an aqueous nitrate solution containing fission products and the like. Not only does this high-level liquid waste exhibit a high calorific value (decay heat) attributed to the decay of radioactive substances and also a high level of radioactivity but also the lifetimes of radioactive substances contained therein are so long that, after safe custody with special attention, it must ultimately be isolated from the zone of human life. Although the high-level liquid waste is now mostly stored in the form of an aqueous solution, it is partly stored in the form of a more stable solidified waste such as glass. This solidified waste as means for disposing of the high-level liquid waste can be safely isolated from the zone of human life for a prolonged period of time by first storing the same for tens of years for cooling and thereafter disposing of it in a stratum as deep as hundreds of meters underground.
The vitrified waste is generally air cooled during its storage so as to prevent its temperature from exceeding the predetermined operating temperature because glass has the property of crystallizing when held at a high temperature for a prolonged period of time to thereby change its characteristics. The cooling capacity depends on the cooling system, i.e., forced cooling or natural cooling of the storage facilities and the cooling capacity design worked out for the storage facilities. Therefore, the upper limit of the waste content in the vitrified waste must be set in accordance with the cooling capacity of the storage facilities so that the maximum temperature of the vitrified waste does not exceed the predetermined operating temperature. The maximum predetermined operating temperature for the currently produced vitrified waste is, for example, about 600.degree. C.
After being stored as described above, the vitrified waste is disposed of in a stratum. Thermal influences on disposal sites, such as thermal stress, heat convection of underground water stream, and deformation of peripheral materials must be minimized.
Most of the radioactivity and calorific value of the high-level liquid waste originate in radioactive isotopes of cesium (Cs) and strontium (Sr), as fission products, and daughter nuclides thereof. For example, with respect to the high-level liquid waste generated in the reprocessing of spent nuclear fuels after a lapse of four years from the takeout from a reactor core at a burnup of 45,000 MWd/tU, the calorific value of Cs and Sr plus Ba and Y that are in radioactive equilibrium therewith accounts for about 65% of the total calorific value. This value approaches 90% after a lapse of 30 years.
Therefore, removing by separation Cs and Sr from the high-level liquid waste followed by solidifying the Cs- and Sr-free liquid waste enables a reduction of the calorific value of the solidified waste, thereby enabling an increase of the waste content of the solidified waste with the result that the waste output can be minimized and a volume reduction can be achieved.
Techniques for separating Cs and Sr from the high-level liquid waste have been proposed which include, for example, one proposed by the same assignee as in the present application in which formic acid is added to the high-level liquid waste to thereby effect denitration [Abstracts of 1993 Spring Meeting of the Atomic Energy Society of Japan, H40 (published on Mar. 10, 1993)]. That is, denitration of the high-level liquid waste with formic acid increases the pH of the liquid waste, and most of the elements other than Cs and Sr (e.g., Mo, Zr, Fe and Y) are precipitated while Cs and Sr remain in the liquid waste by regulating the amount of added formic acid so that the pH of the liquid waste after the denitration fails in the neutral zone of 6 to 7.5 as shown in FIG. 3. Thus, removing the precipitate by separation enables crude separation of exothermic elements consisting of Cs and Sr from the other elements.
However, the actual high-level liquid waste contains tens of varieties of elements and their contents are so widely varied as to provide a complex composition, so that it is difficult to accurately adjust the pH of the liquid waste after the denitration to a given value of the neutral zone in the denitration conducted by adding formic acid. Especially, when the pH of the liquid waste after the denitration exceeds the given range of the neutral zone and is increased to fall on the alkaline side of pH 8 to 9, the problem has arisen that only about 10% of Sr remains in the liquid waste with about 90% thereof being precipitated while Fe and Y are completely precipitated with 90% or more of Cs still remaining in the liquid waste as shown in the graph of FIG. 4, so that the separability of Sr is deteriorated.